Original scientific paper
https://doi.org/10.37798/2025742719
Neutron kinetics modelling for simulations of loss of coolant accidents in the nuclear power plants
Bartlomiej Klis
; Framatome, DTIPD
*
Simon Blaise
; Framatome, DTIPD
Patrick Primeau
; Framatome, DTIPD
Jean-Christophe Lecoy
; Framatome, DTIPD
Marie-Christine Grouhel
; CMAP, CNRS, École Polytechnique, Institut Polytechnique de Paris
* Corresponding author.
Abstract
The code used by Framatome to predict the progression of a LOCA is the system scale thermal-hydraulic code CATHARE (Code Avancé de THermoHydraulique pour les Accidents deRéacteurs à Eau). It calculates the full primary and secondary circuits including the core and the fuel elements. CATHARE currently utilizes a 0D neutronic model that solves the PointKinetics Equation (PKE) to determine the evolution of instantaneous fission power.
This approach is suitable for transients where the core's moderator density changes rapidly in a quasi-uniform manner, such as in large break LOCA scenarios (from the initial situation with a core full of liquid to a sudden and complete voiding leading to a full vapour environment). However, during Intermediate Break (IB) scenarios, with slower dynamics, the assumption of a uniform core's moderator density is no longer valid. This assumption results in a significant underestimation of void antireactivity in the upper part of the core and a slight overestimation at the bottom. Thus, using PKE for IB-LOCA leads to an overestimation of the fission power, and the more heterogeneous the core, the higher the conservatism of this hypothesis is expected.
In fact, the specific application of IB-LOCA involves a precise neutronic calculation in a strongly diphasic fluid environment which is first due to the uncovering of the core and then to its reflooding by the safety injections. The presented work relies on the development of a fine 3D coupling between neutronics and thermal-hydraulics at the assembly scale plunged into a full reactor simulation. Such a development goes beyond the known limitations of current neutronics/thermal-hydraulics couplings (dealing with low void fraction situations) which are not suitable for LOCA safety studies.
Keywords
nuclear safety; thermo-hydraulics; neutronics; code coupling
Hrčak ID:
340889
URI
Publication date:
1.3.2025.
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